1. Field of the Invention
This invention relates to model steam generators of the type used to monitor the corrosion of the heat exchange tubes in nuclear steam generators.
2. Description of the Prior Art
Model steam generators for monitoring the amount of corrosion degradation occurring within the heat exchange tubes of a nuclear steam generator are known in the prior art. Generally speaking, such model generators operate by subjecting an array of sample heat exchange tubes to the same heat, pressure and chemical conditions which surround the heat exchange tubes in nuclear steam generators. If these conditions are accurately simulated, the amount of corrosion which occurs in the sample tubes of the model steam generator will provide an accurate indication of the tube corrosion present in the nuclear steam generator being monitored. Such model steam generators are a particularly useful form of corrosion monitor, because they obviate the need for shutting down the nuclear plant and sending technicians into the radioactive interiors of the generators. However, such model steam generators are useful only insofar as they are capable of accurately simulating the heat, pressure and chemical conditions which exist inside the nuclear plant. Any material departures from these conditions will adversely affect the accuracy of the model steam generator.
In order to understand the difficulties in building a practical model steam generator which provides an accurate monitor for heat exchange tube corrosion, one must first understand how nuclear steam generators are generally constructed, and what chemical and hydraulic conditions are responsible for tube corrosion.
Nuclear steam generators are comprised of three principal parts, including a secondary side and a tubesheet, as well as a primary side which circulates water heated from a nuclear reactor. The secondary side of the generator includes a plurality of U-shaped tubes, as well as an inlet for admitting a flow of feedwater. The inlet and outlet ends of the U-shaped tubes within the secondary side are mounted in the tubesheet which hydraulically separates the primary side of the generator from the secondary side. The primary side in turn includes a divider sheet which hydraulically isolates the inlet ends of the U-shaped tubes from the outlet ends. Hot water flowing from the nuclear reactor is admitted into the section of the primary side containing all of the inlet ends of the U-shaped tubes. This hot water flows through these inlets, up through the tubesheet, and circulates around the U-shaped tubes which extend within the secondary side of the steam generator. The heated water transfers its heat through the walls of the U-shaped tubes to the feedwater flowing through the secondary side of the generator, thereby converting the feedwater to steam. After the nuclear-heated water circulates through the U-shaped tubes, it flows back through the tubesheet, through the outlets of the U-shaped tubes, and into the outlet section of the primary side, where it is recirculated back to the nuclear reactor. The inlet ends of the U-shaped tubes are known as the "hot legs", and the outlet ends of these tubes are known as the "cold legs".
The heat exchange tubes of such nuclear steam generators can suffer a number of different types of corrosion degradation, including denting, stress corrosion cracking, intragranullar attack, and pitting. In situ examination of the tubes within these generators has revealed that most of this corrosion degradation occurs in what are known as the crevice regions of the generator. Such crevice regions include the annular space between the heat exchange tubes and the tubesheet, as well as the annular clearance between these tubes and the various support plates in the secondary side which are used to uniformly space and align these tubes. Corrosive sludge tends to collect within these crevices from the effects of gravity. Moreover, the relatively poor hydraulic circulation of the water in these regions tends to maintain the sludge in these crevices, and to create localized "hot spots" in the tubes adjacent the sludge. The heat radiating from these "hot spots" acts as a powerful catalyst in causing the exterior surface of the heat exchange tubes to chemically combine with the corrosive chemicals in the sludge. While most nuclear steam generators include blow-down systems for periodically sweeping the sludge out of the generator vessel, the sludges in the crevice regions are not easily swept away by the hyraulic currents induced by such systems. Despite the fact that the heat exchange tubes of such nuclear generators are typically formed from corrosion-resistant Inconel stainless steel, the combination of the localized regions of heat and corrosive sludges can ultimately cause the heat exchange tubes to crack, and leak radioactive water from the primary side into the secondary side of the generator. However, this need not occur if the heat exchange tubes are provided with internally reinforcing sleeves before the corrosion causes cracks in the tube walls.
Model steam generators were developed in order to accurately monitor the amount of corrosion degradation occurring in the heat exchange tubes of a particular nuclear steam generator, in order that these tubes might be sleeved before any of the tube walls crack. Such model steam generators have been found to be a particularly accurate way of ascertaining the amount of corrosion degradation occurring in the heat exchange tubes of a nuclear steam generator, because the particular amount of corrosion which the feedwater chemistry and thermohydraulics of the particular generator will induce in a particular set of tubes is virtually impossible to predict by purely theoretical models.
However, such prior art model steam generators are not without significant problems. For example, some of these model steam generators have no means for separating the water out of the wet steam they generate out of their secondary sides. This allows a significant amount of sludge-generating chemicals to escape through water droplets which are entrained within this steam, thereby impairing the ability of the model to accurately simulate the amount of sludge which accumulates in the steam generator. In order to correct this inaccuracy, some model steam generators use scaled-down versions of the swirl-vane type water separators used in full-scale generators.
Such swirl-vane separators operate by rotating a metallic plate through the column of steam generated by the secondary side of the model steam generators. The plate impinges many of the water droplets entrained in this column of steam and slings them against the inner walls of the secondary side of the generator, where they stream back down into the feedwater reservoir in the secondary side. While scaled-down versions of this swirl-vane type separator are effective in removing most of the entrained water droplets in the steam produced in the secondary side of the model generator, the effective separation rate is still much higher than the one quarter of one percent loss rate provided by the swirl-vane type separators in the nuclear steam generators being simulated. Since the scaled-down versions of the swirl-vane type separators are not as proportionally effective at removing entrained water droplets in the steam produced in the secondary side of the model steam generators, these model steam generators have relatively greater water losses through their steam output conduits. These water losses again result in a significant amount of sludge-forming chemicals being injected out of the steam outlet of the secondary side, which again retards sludge formation in the model generator. The retardation of sludge formation throws off the accuracy of the simulation in a particularly dangerous way. The operator of the model steam generator could receive the erroneous impression that the nuclear steam generator being monitored is accumulating less tube-corroding sludge in the vicinity of the tubesheet than it actually is accumulating.
Still other deficiencies in prior art model steam generators include their inability to accurately simulate the condition of the heat exchange tubes and tubesheet in diverse areas of the steam generator. More precisely, many prior art model steam generators are incapable of simulating the circulation and heat flux at various points along the radius of the tubesheet of the nuclear steam generator being monitored. Additionally, many of these prior art generators are incapable of easily simulating both "low flux" and "high flux" conditions which exist in nuclear steam generators from their lowest to their highest steam outputs.
Clearly, a need exists for a model steam generator having a steam separator capable of reducing the water losses through the steam outlet pipe of the secondary side of the generator to a rate commensurate with the water losses which actually occur in the nuclear steam generator being monitored in order to accurately simulate the sludge accumulations which occur in the vicinity of the tubesheet of the generator. Ideally, such a model steam generator should also be capable of accurately and easily simulating the thermohydraulic conditions at any desired point along the radius of the tubesheet of the nuclear steam generator. Finally, such a model generator should have the capacity for simulating the conditions surrounding both the "hot legs" and "cold legs" of the heat exchange tubes over a broad turn-down range.